Constraint-Dependent J-R Curves of a Dissimilar Metal Welded Joint for Connecting Pipe-Nozzle of Nuclear Pressure Vessel

2015 ◽  
Vol 137 (2) ◽  
Author(s):  
J. Wang ◽  
G. Z. Wang ◽  
F. Z. Xuan ◽  
S. T. Tu

In this paper, the J-R curves of two cracks (A508 HAZ crack 2 and A508/Alloy52Mb interface crack 3) located at the weakest region in an Alloy52M dissimilar metal welded joint (DMWJ) for connecting pipe-nozzle of nuclear pressure vessel have been measured by using single edge-notched bend (SENB) specimens with different crack depths a/W (different constraint). Based on the modified T-stress constraint parameter τ*, the equations of constraint-dependent J-R curves for the crack 2 and crack 3 were obtained. The predicted J-R curves using different constraint equations derived from the three pairs of crack growth amount all agree with the experimental J-R curves. The results show that the modified T-stress approach for obtaining constraint-dependent J-R curves of homogeneous materials can also be used for the DMWJs with highly heterogeneous mechanical properties (local strength mismatches) in nuclear power plants. The use of the constraint-dependent J-R curves may increase the accuracy of structural integrity design and assessment for the DMWJs of nuclear pressure vessels.

Author(s):  
K. K. Yoon ◽  
J. B. Hall

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.


Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.


Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.


Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  

Probabilistic fracture mechanics (PFM) has been recognized as a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.


Author(s):  
H. Reece-Barkell ◽  
W. J. J. Vorster

Effective outage planning and implementation is critical to the efficient and safe operation of commercial nuclear power plants in the UK. Statutory outages are necessary for refuelling, for preventive and corrective maintenance when shutdown conditions are required, and for major modification and improvement projects. Outages involve the support of many companies and individuals working together and, as such, require high levels of coordination. Planning of activities before the outage is critical to the overall success of the outage. Establishing the integrity of power plant piping and pressure vessels is a key objective as part of any outage and the methodology and processes involved are the subject of this paper. Establishing the integrity of piping and pressure vessels requires an understanding of the specific threats, their relationship to the overall condition of the system, and the mitigating measures required to assure safe operation. Understanding the specific threats allows the engineering function of an organisation to advise on pipework and pressure vessel ‘Minimum Acceptable Thicknesses’ which can be used to assure integrity via comparison with thicknesses measured during outage inspections. Minimum Acceptable Thicknesses should be recorded in the outage management documentation so they are accessible during the outage implementation phase. Historically a variety of different methodologies have been used to advise on Minimum Acceptable Thickness requirements including design drawing specified minimum thicknesses, design code based required thicknesses and thicknesses calculated based on Fitness for Purpose methods. It is important that a robust procedure be applied to promote consistency of approach as regards the calculation of pipework and pressure vessel Minimum Acceptable Thickness requirements across all power station assets. An additional consideration is that of ensuring that the approach adopted is consistent with high level safety case guidance, i.e., the assessment is appropriate for the failure tolerability of the plant item. This paper provides an overview of the strategy, methodologies and processes employed to determine Minimum Acceptable Thicknesses for pipework components. These ensure that, over a specified inspection interval, were the weld/component to be defect free, it would not fail due to any of the relevant failure mechanisms, which typically are plastic collapse, creep rupture, fatigue, incremental collapse (ratcheting) or buckling. Readers of this paper will gain a valuable insight into the statutory outage process applicable to nuclear power plants in the UK. A particular focus of this paper is on the structural integrity assessments applied in a non-traditional sense prior to, during and after the statutory outage. As well as sharing a valuable insight into the assessment methodologies this paper highlights best industrial practice.


2016 ◽  
Vol 139 (2) ◽  
Author(s):  
Husain J. Al-Gahtani ◽  
Mahmoud Naffa'a

Pressure vessels that undergo repairs are normally pressure tested to verify their structural integrity before returning into service. Conventionally, the entire vessel is pressure tested, according to the relevant construction code. In this paper, partitioning the pressure vessel is suggested as an equivalent alternative test arrangement, where pressure testing is limited to the zone where a repair has been performed. Use of such an arrangement would alleviate potential concerns associated with the conventional testing method. Procedures are provided to specify the position of the partition relative to the repair location, in order to maintain the state-of-stress to that achieved in a conventional pressure test. Validity of this approach has been demonstrated for a repaired full-circumferential welded joint in the wall of a cylindrical pressure vessel.


Author(s):  
Yinsheng Li ◽  
Kunio Hasegawa ◽  
Michiya Sakai ◽  
Shinichi Matsuura ◽  
Naoki Miura

When a crack is detected in a nuclear piping system during in-service inspections, the failure estimation method provided in codes such as the ASME Boiler and Pressure Vessel Code Section XI or JSME Rules on Fitness-for-Service for Nuclear Power Plants can be applied to evaluate the structural integrity of the cracked pipe. In the current codes, the failure estimation method for circumferentially cracked pipes includes bending moment and axial force due to pressure. Torsion moment is not considered. The Working Group on Pipe Flaw Evaluation for the ASME Boiler and Pressure Vessel Code Section XI is developing guidance for combining torsion load within the existing solutions provided in Appendix C for bending and pressure loadings on a pipe. A failure estimation method for circumferentially cracked pipes subjected to general loading conditions including bending moment, internal pressure and torsion moment with general magnitude has been proposed based on analytical investigations on the limit load for cracked pipes. In this study, experimental investigation was conducted to confirm the applicability of the proposed failure estimation method. Experiments were carried out on 8-inch diameter Schedule 80 stainless steel pipes containing a circumferential surface crack. Based on the experimental results, the proposed failure estimation method was confirmed to be applicable to cracked pipes subjected to combined bending and torsion moments.


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